粉末冶金法制备核燃料包壳FeCrAl合金研究进展

燃料包壳是核反应堆安全运行的重要保障。福岛核事故后,国内外开展了大量新型事故容错燃料包壳的研发工作。由于具有抗高温氧化和高强度等优异的综合性能,FeCrAl合金已成为新一代事故容错燃料包壳的重要候选材料之一。经过多年积累,核燃料包壳FeCrAl合金的设计和制备研究已取得一定进展。利用粉末冶金方法制备性能更为优异的氧化物弥散强化FeCrAl合金前景广阔,受到国内外学者的广泛关注。本文综述了核燃料包壳FeCrAl合金的成分设计、熔炼制备和粉末冶金制备的研究现状,分析了不同方法制备合金的组织性能及存在的问题,对未来核燃料包壳FeCrAl合金的设计和制备进行了展望。     

锆合金的织构及其对性能的影响

锆合金被广泛应用于核反应堆中，作为燃料元件的包壳材料或堆内结构部件，织构会影响其众多的应用性能，因此织构的研究及控制在锆合金的开发利用中具有重要作用。综述了，锆合金塑性变形的滑移和孪生系统；锆合金管板材的织构特点及控制方法；热轧温度对锆合金板材织构的影响以及退火处理锆合金织构的演化；并分别总结了织构与锆合金屈服强度、蠕变强度、碘致应力腐蚀开裂、氢化物取向分布以及辐照生长等性能的相互关系。     

钼镧合金表面FeCrAl涂层耐腐蚀性能

采用非平衡磁控溅射工艺在Mo–La合金表面沉积FeCrAl涂层，研究所制备涂层的耐腐蚀性及涂层的腐蚀机理。结果表明：FeCrAl涂层样品在360℃、18.6 MPa、纯水的高压釜中腐蚀72 h，平均腐蚀速率为3.8 mg·dm_?²，低于同条件下锆合金以及未沉积涂层的钼镧合金的腐蚀速率，且涂层中的Al与外界环境介质中的氧发生反应，在涂层表面形成致密的Al₂O₃薄膜，在一定程度上减缓了涂层的腐蚀速度，有效保护了基体材料。FeCrAl涂层样品在1200℃、0.1 MPa的高温水蒸气环境下腐蚀8 h,Al₂O₃氧化膜厚度在4.0μm左右，涂层维持保护效果，钼镧合金基体未暴露在腐蚀环境中；经淬火后，Al₂O₃氧化膜厚度减小至2.5μm左右，涂层依旧维持结构完整性，没有出现贯穿性脱落，满足Mo–La合金表面耐腐蚀性的使用要求。     

高强度耐腐蚀ODS?FeCrAl合金微观结构、力学性能研究进展

氧化物弥散强化（Oxide dispersion strengthened，ODS）FeCrAl合金由于加入一定量的Al元素，使合金表面可形成一层薄而致密的Al₂O₃保护膜，使得合金即便在1400 ℃的水蒸汽下也不会因为腐蚀导致失效。同时，大量超细氧化物粒子的弥散强化作用使其具备优异的高温强度。这种兼具高温强度和耐腐蚀的特性使得ODS?FeCrAl合金成为非常有前景的事故容错燃料（Accident tolerant fuel , ATF）包壳候选材料，也是快堆等其他工作于高温强腐蚀环境的先进反应堆包壳的重要候选材料。Al元素的引入会使ODS铁基合金中弥散粒子的种类发生变化，进而影响其显微组织和力学性能。针对ODS?FeCrAl合金中引入Al元素所导致的显微组织变化及其对蠕变性能的影响，总结了国内外相关研究进展，旨在为适用于先进反应堆的ODS?FeCrAl合金的发展提供参考。      

添加元素对贮氢电极材料 MlNi3.8(CoMn)1.2电化学及高温性能的影响

研究了微量元素Ti,Zr,Al对MlNi3 .8(CoMn) 1.2 合金电化学性能及高温性能的影响。添加少量铝虽然使合金在室温的放电容量和高倍率放电容量降低 ,但可显著提高高温下的容量 ,并有效抑制自放电 ,提高循环稳定性 ;添加少量钛或锆均降低合金的放电容量 ,但可改善高倍率放电性能。锆还可提高合金的循环稳定性 ,延长电极寿命。     

抗高温蒸汽氧化ZrC/Cr复合涂层制备与性能研究

采用预制涂层粉末方式在锆合金表面激光熔覆涂层,成功制备出了不同成分配比的ZrC/Cr复合涂层管。涂层厚度90-120μm,组织均匀致密,与基体结合良好。对涂层管进行高温蒸汽氧化试验,结果发现,ZrC/Cr复合涂层抗高温蒸汽氧化性能受成分配比影响显著,ZrC与Cr粉末配比为1:4的复合涂层在1204℃高温水蒸气环境中表现出了良好的抗高温水蒸气氧化性能,氧化增重约为无涂层锆合金包壳管氧化增重的一半。     

Zr-Sn-Nb-Fe-Cr合金的电子束焊接性能研究

研究了Zr-Sn-Nb-Fe-Cr锆合金真空电子束焊接样品的拉伸和抗腐蚀性能,结果表明,焊接样品室温和375 ℃的拉伸性能略低于母材;焊接样品在500 ℃过热蒸汽中腐蚀500 h未出现疖状腐蚀现象,但腐蚀增重略大于母材;在400 ℃过热蒸汽中腐蚀320 d及360 ℃水中腐蚀520 d,样品焊缝处颜色均为黑亮,未出现白色、棕色等腐蚀产物.根据实验结果拟合出了焊接样品的腐蚀增重曲线.     

酸洗预处理对Zr-4合金高压釜蒸汽腐蚀增重的影响

Zr-4合金和M5锆合金是国内核电站最常用的核反应堆包壳材料,其耐腐蚀性能是一项重要的检测指标。在高压釜蒸汽腐蚀实验中,为了确认腐蚀数据的有效性,每釜应随检测样品放置对比样。针对Zr-4合金作为对比样时的酸洗预处理工艺开展实验,研究是否酸洗及酸洗工艺对其腐蚀增重的影响,以期为建立Zr-4合金对比样规范提供科学的试验方法。研究结果表明,酸洗预处理会明显影响Zr-4合金的腐蚀增重及其标准偏差,未经酸洗处理的Zr-4合金对比样的腐蚀增重较稳定,标准偏差及增重极差值均低于经酸洗预处理的对比样。     

TiN/Cr涂层锆合金包壳高温氧化与抗热冲击性能研究

为了研究TiN/Cr涂层锆合金的高温氧化性能和抗热冲击性能,采用磁控溅射制备了TiN/Cr涂层锆合金样品,分别开展高温氧化和抗热冲击性能试验,并对高温氧化和抗热冲击后的样品微观结构、物相及结合强度等进行表征。结果表明磁控溅射制备的TiN/Cr双层涂层之间存在明显界线,样品表面存在类圆形的团聚凸起,但涂层结构致密,无裂纹和孔隙等缺陷。高温氧化后,表面Cr涂层部分被氧化为Cr₂O₃,呈不规则的多面体结构,中间TiN涂层中Ti原子向表面扩散,与O结合形成TiO₂,呈长条状结构,且Cr₂O₃聚集区域呈鼓泡状。在涂层与基体界面处,向内扩散的Cr原子和向外扩散的Zr原子形成约5μm厚的Cr-Zr扩散层。同时,O原子持续向样品内部扩散,与基体中Zr原子结合形成ZrO₂。虽然TiN/Cr涂层锆合金的氧化增重小于无涂层锆合金,且样品保持相对完整,但是由于O原子的持续扩散,内部锆合金被氧化,说明TiN/Cr涂层无法为锆合金提供良好的长时抗高温氧化性。热冲击试验后,表面涂层依然...     

SiO2浓度对Zr-4合金包壳管力学和腐蚀性能的影响

Zr-4合金包壳管在核工业中应用广泛,为探究一回路水质环境中SiO2的浓度对Zr-4合金组织性能的影响,利用高温高压反应釜模拟锆管在反应堆一回路的工作环境来开展实验.以显微硬度值和腐蚀失重值为指标,研究了锆管在不同SiO2浓度下的力学性能和腐蚀性能.结果表明,随着SiO2浓度的增加,Zr-4合金的显微硬度值逐渐提高,但在腐蚀介质环境中的耐腐蚀性能明显降低.     

Research progress of nuclear FeCrAl cladding materials

The leakage accident at the Fukushima nuclear power plant in Japan exposed the defect that zirconium alloy easily reacts with water vapor to cause an explosion, and distracting people's attention back to the FeCrAl alloy, which has relatively balanced properties. FeCrAl alloy has better corrosion resistance than zirconium alloy, and can better meet the working conditions of nuclear reactor under high temperature water vapor environment. In order to further ensure the safety of nuclear reactors, scholars from various countries had conducted a lot of research on the performance of FeCrAl alloy. For that reason, the corrosion mechanism of FeCrAl alloy and its corrosion behavior in high temperature steam and high temperature water environment were reviewed, and the effect of alloying elements (rare earth, Ti, Zr, Nb, and Mo), deformation and heat treatment system on the mechanical properties of the alloy was discussed; the research progress of the radiation and welding performance of FeCrAl alloy was summarized. The deficiencies in the research of nuclear grade FeCrAl alloys and the key development directions in the future were pointed out, which would provide references for other scholars in the field of research in the future.     

固溶热处理对新型核用FeCrAl合金包壳管组织和力学性能的影响

通过新型核用锻态FeCrAl合金包壳管在不同温度及时间下的固溶处理实验,利用光学显微镜(OM)、扫描电子显微镜(SEM)、X射线衍射(XRD)、维氏硬度仪(HV)和电子万能试验机(EUTM)分析了固溶处理对FeCrAl合金包壳管显微组织以及力学性能的影响规律。结果表明,当固溶热处理温度低于1200℃时,保温60 min后FeCrAl合金组织中仍然存在一定数量的σ-FeCr相、富Cr相和Fe₂(Mo/Nb)Laves析出相。随着温度的升高,析出相逐渐溶解到基体之中,当温度升高到1200℃时,保温60 min后组织内的析出相基本回溶,但晶粒发生了异常长大。随着保温时间的不同,析出相的数量、形态、分布和成分都会发生变化。随着温度的升高,FeCrAl合金的抗拉强度逐渐降低,延伸率先升高再降低,硬度先减小再增大。在1150℃加热、60 min保温和水冷的固溶热处理制度下,核用FeCrAl合金包壳管的伸长率可达20.1%,抗拉强度R_m为675 MPa,屈服强度R_p0.2为560 MPa,硬度为HV 261,塑性提高,加工性得到显著改善,可满足管材后续轧制的要求。     

Corrosion of Zirconium Alloys in Nuclear Applications – A Review

With the recent demands for higher fuel discharge burn up towards better fuel utilisation and higher heat ratings and coolant temperatures for better thermal efficiency, interest on corrosion of zirconium alloys has been rekindled even though a large amount of work has been carried out to understand various aspects of corrosion of zirconium and its alloys in relation to their behaviour under nuclear operating conditions in the last 40 years or so. Though it seems that there exists some scope for further development of these alloys, which may need better understanding of the various corrosion mechanisms/micromechanisms, in the present paper an attempt has been made to highlight a few corrosion aspects related to zirconium, however mainly zircaloys. Investigations carried out in this laboratory as well as elsewhere have been discussed keeping in view the aim of advancing the present knowledge so as to be able to meet the future requirements in terms of corrosion resistance as expected from this alloy system.     

Dynamic Precipitation of Laves Phase in FeCrAl Alloy with Zr Addition

FeCrAl alloy is one of potential candidates for accident-tolerant-fuel (ATF)-cladding materials due to its excellent oxidation and corrosion resistance at accident temperature, combining good mechanical properties at service temperature. Alloying strategy is an important way for improving comprehensive properties of FeCrAl alloy through the precipitation of fine Laves phase. Zr alloying can stabilize the Laves phase due to its lower diffusion coefficient and solubility in body-centered-cubic ferrite matrix. Herein, it is found that Zr addition changes the dynamic precipitation features of Laves phase in FeCrAl alloy during high-temperature deformation, from only one type of Fe₂M (M = Nb, Mo, Ta) Laves phase to Fe₂Zr combining Fe₂M-type Laves phase. The Fe₂Zr-type Laves phase precipitates dynamically first, and the interface precipitates between which with ferrite matrix creates more nucleation sites for subsequent precipitation of Fe₂M Laves phase. The results can be possibly applied for alloy design and microstructure tailoring in series of FeCrAl alloys used for ATF cladding in the near future.     

Corrosion of Zirconium Alloys in Nuclear Applications - A Review

Abstract

With the recent demands for higher fuel discharge burn up towards better fuel utilisation and higher heat ratings and coolant temperatures for better thermal efficiency, interest on corrosion of zirconium alloys has been rekindled even though a large amount of work has been carried out to understand various aspects of corrosion of zirconium and its alloys in relation to their behaviour under nuclear operating conditions in the last 40 years or so. Though it seems that there exists some scope for further development of these alloys, which may need better understanding of the various corrosion mechanisms/ micromechanisms, in the present paper an attempt has been made to highlight a few corrosion aspects related to zirconium, however mainly zircaloys. Investigations carried out in this laboratory as well as elsewhere have been discussed keeping in view the aim of advancing the present knowledge so as to be able to meet the future requirements in terms of corrosion resistance as expected from this alloy system.     

事故容错燃料包壳用Mo合金的研究进展

2011年发生福岛严重核事故后,锆合金包壳材料的安全可靠性受到严重的质疑,国内外对事故容错燃料(ATF)开始了广泛的研究。Mo合金由于其优异的高温性能成为了ATF的候选包壳材料之一。本文综述了Mo合金包壳材料在高温氧化性能,力学性能,抗辐照性能,中子经济性能以及加工和焊接性能方面的研究进展,并指出在工业应用中面临的挑战,最后展望了Mo合金在ATF包壳材料中的应用前景。     

锆合金变形机理及其板材织构演化规律

锆是核工业的重要结构材料,又是优秀的化工耐蚀结构材料。锆合金的织构会对它的屈服强度、蠕变和疲劳强度、应力腐蚀开裂行为以及辐照尺寸变化等产生很大影响,因此变形机理的研究和织构控制在锆合金的开发利用中有十分重要的地位。综述了锆合金的变形机理,介绍了锆合金板材在不同轧制温度下的织构演化规律,以及退火温度对锆合金板材织构的影响,并总结了织构对锆合金板材力学性能的影响。最后指出,目前对锆合金板材加工后的织构进行精确预测还十分困难,需进行详细深入的研究,同时在加工中产生的织构对加_丁过程的影响以及与温度、应力分布、合金成分和组织的关系还需进一步认识。